ML19322C582
| ML19322C582 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Crane  |
| Issue date: | 09/04/1979 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19322C578 | List: |
| References | |
| TASK-TF, TASK-TMR NUDOCS 8001170907 | |
| Download: ML19322C582 (3) | |
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4 UNITED STATES
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,8 September 4, 1979
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Dockets Nos. 50-269/270/287 MEMORANDUM FOR:
Domenic B. Vassallo, Acting Director, Division of Project Management FROM:
Darrell G. Eisenhut, Acting Director, Division of Operating Reactors
SUBJECT:
DOR RESPONSE TO NRC/TMI SPECIAL II ;UIRY GROUP (SIG) REQUEST FOR OCONEE STATION OPERATIONAL.ISTORY This memorandum is the DDR pcrtion of the response to the request for infor-mation from DeYoung, SIG, to Denton/Stello dated July 24, 1979, enclosed.
DDR,.after discussion with IE Region II, is responding to those items transferred to D0R by IE referred to in the second part of Item 3 of the SIG request.
OPM: took responsibility for Items 1 and 2, IE Region II, is responding to the first part of Item 3.
The second part of Item 3 asked for a description of the staff's handling of "significant" events and how the lessons learned from the events were constructivel,.
used to prevent future adverse consequences.
All significant events at an operating plant are normally reported to IE through a Licensee Everst Report.
IE will transfer those events that require licensing action to resolve, and these generally require additional review by DOR.
Only three such significant events at Oconee were transferred to DOR:
Possible Use e.f Atypical Weld Wire in Reactor Vessel Welds, Turbine Building Flooding and Steam Generator Tube Failures, g
,r We were informed of the atypical weld wire problem brough a Part 21 submittal h.
by Babcock & Wilcox to the NRC dated August 4,19 Duke contacted IE the same day to report that weld material in the Oconee 3 reactor vessel may be different from the Mill Certifications.
IE transferred the review to DOR.
We were informed that as many as 12 reactor vessels could be involved that were manufactured by 5&W, four in Westinghouse systems, seven in B&W systems and one GE vessel.
B&W conducted an investigation of QA records at their plant to determine which vessels had atypical material. Tne investigation was inconclusive.
On August 14, 1978 a
- er.eric lettsr was prepared asking all potentially affected plants to put more
- 0nserva-ive heatup and cooldown curves into use.
We had oreviously phoned
.nese ;iants.
By-August 23, 1978 all the licensees had responded that B&L', or e ".EES if a nor. ESW syster, had su: plied the more c nservative curves and L.at a'
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3 they were in use.
B&W completed an in-reactor irradiation at Crystal River Unit 3 of the atypical material and submitted a preliminary report to NRC.
We have concluded that continued operation of these plants is acceptable with the more conservative heatup and cooldown curves in use.
We were informed of the Turbine Building flooding by Licensee Event Report R0-287/76-18 dated October 25, 1976 addressed to 0IE, Region II.
The event occurred during a main condenser inspection.
The discharge water surface is
. at a higher level than the Turbine Building floor, the condenser manways were opened and a condenser circulating water discharge valve failed open, which resulted in.a flooding path.
The FSARhad assumed
?. flood of about 1000 cfs from the condenser circulating intake pipe, w:ich was greater than the actual discharge. side flood.
The licensee propos:d installing protective walls around vital equipment in the Turbine Building and separating the Turbine Building from the Auxil.iary Building by waterproofing and sealing the common wall between the buildings.
Duke's proposal was submitted in a letter dated April 21,1977; the proposals were first discussed with the ONRR staff during, two meetings in November 1976.
Subsequent to Duke's April 21, 1977 letter the modifications were performed under 10 CFR 50.59(a)(1).
Duke, in order to reduce th6 number of vital areas in their Station Security Plan proposed a Safe Shutdown Facility independent of the present shutdown capability.
This Safe Shutdown Facility would also serve to get t.he plant in a safe configuration after either a flooding event or a fire.
This facility is currently under
.0 construction and should be operable by the end of 1980.
This flooding event at Oconee was unique in that the plant had a heat sink water surface at a higher elevation than the Turbine Building floor level.
There is an ongoing review of all plants that started as a result of this flood in addition to the generic flood review c'aused by the event at Quad Cities.
The Oconee Unit No.1 steam generators suffered recurring leaks that raised quest' l
over continued safe operation of B&W once-through steam generators.
One primary concern was how many simultaneous tube failures could be tolerated, say in the event of a main steam line break, and not exceed Part 100 doses at the site boundary.-
A' series of seven Licensee Event Reports dated between October 31, 1976 and April 27, 1978 were submitted by Duke describing 10 tube leaks, tube inspections and tube removal or plugging operations.
The first leak reported in October 31, 1976 occurred in SG 1A, the remaining nine leaks occurred in SG 1B.
Since April 27, 1978 another LER was submitted by Duke for a SG 1B leak, the LER was dated August 20, 1979.
The DDR staff held many meetings with Duke and B&W, sent many formal requests for information and prepared a Safety Evaluation dated October 4,1977.
This SE effected a reduction in the primary to secondary Technical Specification leak limit through a SG tube from 1.0 gpm to 0.3 cpm and found that we understood the mechanism of degradation and rate c' ce:rada:icn so that the SG could continue to operate tnroughout the inspection
- i. erval.
- une submitted a Safety Analysis dated September 9,1977, which i,dicated that up to ten SG tubes could undergo a double ended rupture and
- n:le:e separation of the encs during a main steam line break and that the cor.:
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r quences would not exceed Part 100. The staff has not completed the evaluation of this submittal. Our review indicated two separate degradation rechanisms, one an erosion / corrosion effect at the 14th support plate level and lane tube degradation.
The review resulted in Technical Specification changes for B&W operating plants in addition to the reduced tube leak limit a' Oconee 1.
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D. G. Eisenhut, Acting Director Division of Operating Reactors Office of Nuclear Reactor Regulation
Enclosure:
Memo.to HRDenton & VStello
'fm. P.DeYoung dtd. 7/24/79 re: Request for Information cc:
RVollmer BGrimes F
WGammill LShao JRMiller TJCarter WRussell RReid MFairtile RIngram VNoonan b
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